Sweden has 10 nuclear power reactors providing nearly half of its electricity. Due to the
current concern
for optimal energy sources, many Swedes consider nuclear power a good option when
competitiveness and environmental impact are taken into account. As nuclear power carries the
risk of toxic pollution, the safe, efficient operation and maintenance of the
nuclear reactors is very important.

Because of its reliability and solution effectiveness, ADINA is used in many reactor studies in Sweden.
In this News we feature one such study of the effect of a pipe break in a pressurized water
reactor. The study was conducted by Onsala Ingenjörsbyrå (Onsala Engineering AB, Sweden)
which performed the work under a contract from Ringhals AB. Figure 1 below shows
the nuclear reactor considered.

Fig. 1. Pressurized water reactor studied

The geometry for the entire model was created in Pro/Engineer, while the mesh and the
boundary condition sets were created in ANSA. The complete model consists of more than
800,000 solid, structural and fluid elements with 55,000 nodes in contact, a total of 1.6 million
degrees of freedom. The model is depicted in Figure 2 below, and part of the mesh
used to model the reactor vessel is shown in Figure 3.

Fig. 2. Finite element model of reactor

Fig. 3. Part of the mesh consisting of solid, structural and fluid elementsused to model the reactor vessel

The model was originally intended to calculate the loads, due
to pipe breaks in the system, on the bolts holding in place the
baffle plates that are used to direct the flow of coolant within the reactor vessel.
The model was later extended to include the entire primary system in order to calculate
the response of the whole system to pipe breaks.

The finite element model was imported to the AUI via the Nastran format, then completed
in the AUI for subsequent solving by ADINA. Finally, the AUI was used to post-process
the results.

The above movie shows the pressure distribution in the reactor vessel due to a pipe
break in the cold leg of the reactor coolant loop (depicted by the breaking gray line
at the left side of the movie). Figure 4 below shows the pressure distribution at
a cross section of the reactor vessel just after the break.

Fig. 4. Pressure plot in section of reactor vessel

With the results of the analysis, Onsala engineers were able to predict the structural response
of the reactor vessel due to the pipe break (Figure 5 below) as well as the maximum
bolt forces.

Fig. 5. Predicted plastic regions in reactor vessel

While already yielding many useful results, of course, the model can be
further developed for additional analyses.

This study illustrates, only to some degree, the powerful capabilities
available in ADINA for dynamic linear and nonlinear analyses, including
fluid-structure interactions. Many different types of analyses can be
performed in a very effective and reliable manner, a requirement
particularly important in studies of reactors.